Zircaloy tubing having high resistance to crack propagation

Abstract: A Zircaloy cladding (104) having an outer region comprising fine precipitates (108) and inner region comprising coarse precipitates (102) is provided. The outer region comprises about 10% and the inner region comprises about 90% of the cladding wall thickness. Such Zircaloy tubing is resistant to propagation of cracks and at the same time resistant to corrosion in boiling water reactors (BWR). Resistance to damage caused by the pellet-cladding-interaction can be achieved by standard application of a zirconium or zirconium-alloy liner on the tubing inside surface.

Description: This invention relates to Zircaloy cladding for use in nuclear fuel elements. More particularly, the invention relates to cladding having improved axial crack propagation resistance. BACKGROUND OF THE INVENTION Nuclear reactors have their fuel contained in sealed cladding for the isolation of the nuclear fuel from the moderator/coolant system. The term cladding, as used herein, refers to a zirconium based alloy tube composed of at least one metal in addition to the zirconium base. The term precipitates, as used herein, refers to added metals of the cladding and forming isolated structures in a matrix throughout the zirconium alloy. These precipitates may or may not constitute intermetallics. Typically, these precipitates are uniformly distributed in the matrix -- although they vary in size. Further, so-called fine precipitates (below 0.1 µm (microns)), can either be in the matrix format or the so-called two dimensional format where the precipitates occupy sheet like layer near the outer surface of the zirconium alloy. The cladding -- nominally in the order of 0.076cm (0.030 inches) thick -- is formed in the shape of a tube with the nuclear fuel contained typically in pellet form therein. These pellets are stacked in contact with one another for almost the entire length of each cladding tube, which cladding tube is in the order of 406cm (160 inches) in length. Typically, the cladding tube is provided with springs for centering the fuel pellets and so-called "getters" for absorbing fission gases. Thereafter, the internal portions of the fuel rod are pressurized with various gases for optimum dissipation of gases produced from the fission reaction, and sealed at both ends. Zirconium and its alloys, under normal circumstances, are excellent nuclear fuel cladding since they have low neutron absorption cross sections and at temperatures below about 398° C (at or below the core temperature of the operating reactor) are strong, ductile, extremely stable and nonreactive in the presence of demineralized water or steam. "Zircaloys" are a widely used family of corrosion-resistant zirconium alloy cladding materials. The Zircaloys are composed of 98-99% by weight zirconium, with the balance being tin, iron, chromium, and nickel. "Zircaloy-2" and "Zircdloy-4" are two widely-used zirconium-based alloys for cladding. (Zircaloy-4 omitting nickel). Cladding corrosion is a potential problem both in boiling water reactors and pressurized water reactors. For example, in a PWR, water does not boil -- although in modern designs minute boil can occur at the top of some fuel rods. The oxygen level is relatively suppressed, being about 20 ppb. Hydrogen is injected and resident in the water moderator at about 200 ppb and utilized to suppress oxygen levels. Water pressure is in the range of 140.6 kg / cm 2 (2000 psi) with temperature ranging from 300°C to 380°C dependant upon the operating state of the reactor. Corrosion in PWR cladding is uniform and related to precipitate size in the Zircaloy cladding. Small precipitates have been found to actually accelerate the uniform corrosion phenomena. Consequently, relative large precipitate sizes are preferred in the PWR zirconium cladding. In the radiation environment within the PWR, the precipitates dissolve and become smaller with exposure. To avoid accelerated uniform corrosion buildup, PWR cladding uniformly starts with large precipitate sizes -- .2 microns and above -- to slow the formation of small size precipitates and the more rapid uniform corrosion that occurs with the small size precipitates. In a BWR environment, water does boil. The oxygen level is relatively high, being about 200 ppb. Hydrogen may be injected for the stability of structural parts of the reactor, is effectively stripped off as a part of the boiling, and is resident in the water moderator in the range of 20 ppb. Water pressure is in the range of 70.3 kg / cm 2 (1000 psi) with temperature at 288°C being essentially a function of pressure and for the most part constant all operating rates of the reactor. Corrosion in a BWR occurs in nodular or pustule formats on the zirconium cladding. Uniform corrosion is also present -- but in the usual case not to a significant degree. Further, mineral and particle deposition occurs on the water exposed surface of the cladding. The combination of the corrosion and depositions can become fairly thick on the water exposed portions of the cladding. Nodular or pustule corrosion is not inherently bad. However, where fuel in the reactor has longer life -- such as time within the reactor exceeding 40 megawatt days per ton, nodular or pustule corrosion concentrates. Where such nodular or pustule corrosion becomes concentrated and acts in conjunction with other particles -- such as copper ions -- localized penetration of the cladding wall can occur. Small precipitates have been found to actually suppress nodule and pustule formation. Consequently, it is desired to have small precipitates -- below .1 microns -- to inhibit formation of nodules or pustules. It is known in the prior art to externally treat the outer water exposed surface of cladding with heating from a coil to produce a fine precipitate exterior surface. See Eddens et al. U.S. Patent 4,576,654. In the radiation environment within the BWR, the precipitates dissolve and become smaller with radiation exposure. Nodular corrosion is inhibited by the small precipitates and by the alloying elements put in solution by the dissolution process. Anneals of zirconium alloys have been used and can be summarized in terms of temperature ranges. Starting at low temperatures, anneals above 480°C effect stress relief, usually after working of the metal to achieve around 70% reduction in area. Anneals at about 576°C not only effect stress relief but also commence recrystallization of the metal. In such anneals, maximum ductility of the material is achieved. Finally, anneals substantially above 576°C effect crystal growth -- generally softening the metal. In the prior art, the heat treatment for PWR cladding has included high temperature anneals with slow quenches (less than 5°/second) to preserve large precipitate sizes. Conversely, the heat treatment for BWR cladding has included low temperature anneals with fast quenches (greater than 5°/second) to produce small precipitate sizes. The corrosion resistance of Zircaloy cladding has been improved by forming small, uniformly distributed precipitate particles in the Zircaloy metal matrix. Some portion of the iron, chrome, and nickel components in the Zircaloy matrix form insoluble crystalline precipitates having chemical compositions distinct from the matrix. The precipitates are generally represented by the chemical formulas Zr(Fe,Cr) 2 and Zr 2 (Fe,Ni). Typically the precipitates used in the more corrosion resistant alloys have an average diameter of less than about 0.1 microns. Corrosion and cracking can both damage cladding, but they are fundamentally different phenomena. Cracking is a mechanical breaking or splitting of the cladding wall, while corrosion is an electrochemical conversion of the cladding metal into an oxide or other non-metallic compound. Cracks may be initiated by a variety of causes including mechanical stresses as well as corrosion. Once a crack is initiated, it may pose little problem, so long as it remains confined to a small area. However, if the crack propagates, the cladding can be breached and the fission material eventually contacts the coolant or moderator. Ultimately, this can lead to an expensive reactor outage. The mechanical initiation of cracks can be attributed to various stresses in a conventional reactor. Cracks can start when debris such as wires or metallic shavings or particles find their way into reactor water that flows within the fuel bundles between the fuel rods. The debris may lodge at a fuel rod spacer adjacent the cladding wall. As a result, the debris vibrates or frets against the cladding wall under the influence of the passing steam/water mixture. Such vibration continues until a crack begins. Corrosion can be the source of initial crack propagation. Moreover, manufacturing defects can be the points of crack origin. Still further, crack propagation can start on the inside of the fuel rods in the corrosive high pressure environment present during in service reactor life. Regarding cracking in the interior of the sealed cladding tube, brittle splitting of such cladding may occur due to the combined interactions between the nuclear fuel, the cladding, and the fission products produced during the nuclear reaction. It has been found that this undesirable performance is due to localized mechanical stresses on the fuel cladding resulting from differential expansion and friction between the fuel and the cladding. These localized stresses and strain in the presence of specific fission products, such as iodine and cadmium, are capable of producing cladding failures by phenomena known as stress corrosion cracking and liquid metal embrittlement. Other phenomena such as local hydriding of the cladding and the presence of oxygen, nitrogen, carbon monoxide, and carbon dioxide can assist cladding failure and lead to rod cracking. U.S. Patents 4,200,492 and 4,372,817 to Armijo et al as well as Adamson U.S. Patent 4,894,203 suggest solutions to preventing crack initiation by including a barrier on the inside of the cladding. Cladding containing introduce barrier are sometimes referred to as "composite" cladding or cladding having two distinct metallurgical layers. EP-A-0,425,465 discloses a method of producing a zirconium alloy cladding tube which exhibits improved nodular corrosion resistance while retaining strength and ductility. Although it is highly desirable to prevent crack initiation, in the event a crack forms, its propagation is to be avoided. There exists a need, especially for a BWR environment, for cladding which is resistant to axial crack propagation. There also exists a need for cladding which, iii combination, is resistant to axial crack propagation, crack initiation and corrosion. SUMMARY OF THE INVENTION According to the invention, there is provided a zirconium alloy cladding tube with a metallurgical gradient, said cladding tube comprising a corrosion resistant outer circumferential tube region comprising relatively fine and more densely distributed precipitates and a crack resistant interior circum- ferential tube region CHARACTERIZED BY said interior circumferential tube region comprising coarser and less densely distributed precipitates having an average diameter of at least 0.15 µm (microns) produced by coarse anneal at temperature of greater than 621°C. The coarse precipitates impart resistance to propagation of cracks (especially in the axial direction) and the fine precipitates impart resistance to nodular corrosion in nuclear reactors. Further, resistance to fuel rod container damage caused by the pellet-fuel rod interaction can be achieved by utilizing a zirconium or dilute zirconium-alloy liner on the tubing inside surface. Because nodular corrosion is most problematic on the outer surfaces of typical zirconium alloy tubing in boiling water reactors (as well as many other applications), and in one embodiment, the fine precipitates are formed only on the outer regions of the tubing. The bulk of the tube contains coarse precipitates to provide improved resistance to crack propagation. The transition between the fine precipitate outer region and the coarse precipitate inner region will exist as a relatively sharp boundary. Further details and embodiments of the invention are provided in the following discussion and associated drawings. BRIEF DESCRIPTION OF THE DRAWINGS Fig. 1 is cross-sectional view of a preferred fuel rod of the present invention; Fig. 2 is a partial cutaway perspective view of a nuclear fuel bundle containing nuclear fuel rod; Fig. 3 is a partial cutaway sectional view of a nuclear fuel bundle showing the interior of a fuel rod; and Fig. 4 is a schematic diagram showing regions of fine and coarse precipitates in tubing produced according to the present invention; and Fig. 5 is a cross-sectional view of a second preferred fuel rod in accordance with the invention. DESCRIPTION OF THE SPECIFIC ZMBODIMENTS I. THE TUBING STRUCTURE As used herein, the term "section" refers to a piece of metal having various uses and shapes. The term "tubing" refers to a metal tube having various uses, and the term "fuel rod container" or simply "container" refers to tubing used in fuel rods to enclose fuel pellets. Sometimes the fuel rod container is referred to as "cladding". Referring to Fig. 1, a fuel element 14 (commonly referred to as a fuel rod) includes a fuel rod container 17 surrounding a fuel material core 16. The fuel element 14 is designed to provide excellent thermal contact between the fuel rod container 17 and the fuel material core 16, a minimum of parasitic neutron absorption, and resistance to bowing and vibration which is occasionally caused by flow of coolant at high velocity. The fuel material core is typically a plurality of fuel pellets of fissionable and/or fertile material. The fuel core may have various shapes, such as cylindrical pellets, spheres, or small particles. Various nuclear fuels may be used, including uranium compounds, thorium compounds and mixtures thereof. A preferred fuel is uranium dioxide or a mixture comprising uranium dioxide and plutonium dioxide. The fuel rod container 17 of this invention has an inner region 30 containing coarse precipitates and an outer region 33 containing fine precipitates. Both the inner and outer regions of fuel rod container 17 are preferably comprised of a zirconium alloy such as Zircaloy-2 or Zircaloy-4. Zircaloy-2 has on a weight basis about 1.5 percent tin; 0.12 percent iron; 0.09 percent chromium and 0.05 percent nickel. Zircaloy-4 has essentially no nickel and about 0.2% iron but is otherwise substantially similar to Zircaloy-2. "Zirlo" -- a zirconium based alloy containing about 1% tin, about 1% niobium, and less than about 0.2% iron -- is another alloy that may be used with this invention. Still other alloys that may be used with present invention include at least about 98% zirconium, between about 0.06 and 0.25% iron, between about 0.03 and 0.1% nickel, and between about 0.8 and 1.7% tin (all percents by weight). Other additives may include niobium, bismuth, molybdenum, as well as various other elements used in the art. Most generally, any zirconium alloy that forms intermetallics may be employed. Fig. 2 shows a cutaway sectional view of a nuclear fuel bundle or assembly 10 which is a discrete unit of fuel containing many individual sealed fuel elements or rods R. The fuel assembly consists of a flow channel C provided at its upper end with an upper lifting bale 12 and at its lower end with a nose piece L and lower lifting bale 11. The upper end of channel C is open at 13 and the lower end of the nose piece is provided with coolant flow openings. The array of fuel elements or rods R is enclosed in channel C and supported therein by means of upper tie plate U and lower tie plate (not shown). Certain of the fuel rods serving to "tie" the tie plates together -- thus frequently being called "tie rods" (not shown). In addition, one or more spacers S may be disposed within the flow channel to hold the fuel elements in alignment with one another and the flow channel. During the in service life of the fuel bundle, the liquid coolant ordinarily enters through the openings in the lower end of the nose piece, passes upwardly around fuel elements R, and discharges at upper outlet 13 in partially vaporized condition. Referring now to Fig. 3, the fuel elements or rods R are sealed at their ends by end plugs 18 welded to the fuel rod container 17, which may include studs 19 to facilitate the mounting of the fuel element in the fuel assembly. A void space or plenum 20 is provided at one end of the element to permit longitudinal expansion of the fuel material 16 and accumulation of gases released by the fuel material. A getter (not shown) is typically employed to remove various deleterious gases and other products of the fission reaction. A nuclear fuel material retainer 24 in the form of a helical member is positioned within space 20 to provide restraint against axial movement of the pellet column during handling and transportation of the fuel element. Fig. 4 depicts the microstructure of a tube wall 104 according to this invention. As shown in the outer region of the tube wall, the zirconium alloy exists as grains 100. In the inner region of the tube wall, the alloy exists as grains 106. Generally, the grains in the inner and outer regions will have roughly the same size. The microstructure of the grains, however, is different in the two regions. As shown, the outer region grains contain dense collections of fine precipitates 108, while the inner region grains contain less dense collections of coarse precipitates 102. In some embodiments, the inner surface of tube wall 104 will contain a zirconium liner or barrier. Zircaloy regions containing fine precipitates are more resistant to nodular corrosion than regions containing coarse precipitates. In the present invention, the precipitate structure is altered such that at the outer surface of the tubing wall which is typically exposed to the corrosive boiling water environment, the "fine" precipitate is present while the more "coarse" precipitate is present on the inner surface and throughout the bulk of the tubing wall. The coarse precipitates are believed to impart resistance to crack propagation, especially along the axial direction in a tube. Preferably, the fine precipitates employed in this invention have an average diameter of between about 0.01 and 0.15µm (microns), and more preferably between about 0.02 and 0.06µm (microns). Most preferably, the average diameter of the fine precipitates is about .04µm (micron). Preferably, the coarse precipitates have an average diameter of between about 0.15 and 2.µm (microns), and more preferably between about 0.2 and 1µm (microns). The above precipitate diameters are easily determined by standard transmission electron microscopy (TEM and STEM) known in the art. It is recognized that precipitates may be distributed throughout the Zircaloy matrix uniformly or as a two dimensional array. In this invention, the fine precipitates in the outer region may be distributed in either manner. The coarse precipitates, in contrast, generally will be distributed uniformly only. Although corrosion is normally most pronounced at the tubing surfaces, crack initiation and crack propagation occur throughout the entire tubing wall. Thus, the tubing of this invention preferably will have coarse precipitates present throughout the bulk of the tubing, while having fine precipitates present only near the outer surface of the tubing. In preferred embodiments, the outer region containing the "fine" precipitate will span between about 5 and 20% of the wall thickness, while the inner or bulk region containing the coarse precipitate will span between about 80 and 95% of the wall thickness. In more preferred embodiments, the outer region will span between about 5 and 15%, and most preferably, about 10% of the wall thickness. Thus, in typical tubing having a wall thickness of 0.076cm (30 mils), the outer region will preferably occupy about 0.0076cm (3 mils). While not wishing to be bound by the following theory, it is believed that coarse precipitates impart resistance to crack propagation by virtue of their larger size as well as the effect they have on the local concentrations of iron and nickel at the precipitate/matrix interface. With respect to Zircaloy-2, although the ratio zirconium, nickel, iron, and chrome remains fixed in the precipitates, it can vary in the Zircaloy matrix, especially near the precipitate/matrix interface. As the precipitate size decreases, less of the iron and nickel remains in the insoluble precipitates. Thus, Zircaloys having fine precipitates also tend to have increased concentrations, albeit relatively low concentrations, of iron and nickel in the matrix phase near the precipitates. These two effects (small precipitate size and increased solute concentration) are believed to contribute to reduced resistance to crack propagation. Low solute concentrations in the Zircaloy matrix tend to impart increased ductility. Thus, as a crack point encounters a precipitate/matrix interface, it is believed that it will propagate slower (or even stop) in regions of lower solute concentration because these areas are more ductile. The precipitate characteristics discussed thus far refer to the precipitates in the tubing immediately after formation, but before installation in a reactor. It is well known that during reactor service neutron irradiation changes the precipitate characteristics. These changes include (1) amorphization of the originally crystalline Zr(Fe,Cr) 2 precipitates; (2) transfer of iron from the Zr(Fe,Cr) 2 precipitate to the Zircaloy matrix as solute; (3) transfer of nickel from the Zr 2 (Fe,Ni) precipitates (in Zircaloy-2) to the Zircaloy matrix as solute; and (4) decrease in precipitate size. In some cases, so much material is driven from the precipitates that the surrounding Zircaloy matrix becomes locally supersaturated with the iron and nickel. All of these changes are in the direction of improving nodular corrosion resistance but decreasing crack propagation resistance. The rate of these changes is a function of the neutron fluence (time in the reactor) and the original precipitate size. The changes are more rapid for small precipitates than for large precipitates. Therefore by maximizing the number of large precipitates in the bulk of the Zircaloy tubing, the degradation of the tubing (in terms of resistance to crack propagation) due to neutron fluence is improved. II. TUBING CONTAINING BARRIER LAYERS In some preferred embodiments, barrier layers having low neutron absorption are formed on the tubing inner surfaces to resist tubing degradation caused by interaction with the nuclear fuel. It is known that the fuel rod containers can be split or otherwise degraded due to interactions between the nuclear fuel, the fuel rod container, and the fission products of the fuel. The barrier layer is typically a highly pure zirconium (such as crystal bar zirconium) or moderately pure zirconium (such as sponge zirconium) sheath metallurgically bonded to the inner surface of the tubing. See U.S. patent Nos. 4,200,492 and 4,372,817 to Armijo and Coffin , U.S. patent No. 4,610,842 to Vannesjo , and U.S. patent No. 4,894,203 to Adamson . Cladding containing such barrier layer is shown in Figure 5. In addition to the various elements presented in the embodiment of Figure 1, the fuel rod container of Figure 5 includes a barrier layer 31 bonded to the inner surface of inner region 30. A typical barrier layer (which will form a third component of the structures of this invention) will form about 1 to 30 percent, and more preferably about 5-15%, of the total tubing thickness. The barrier layer should be affixed to the remainder of the tubing by a strong bond. There should be sufficient diffusion between the materials of the tube and the barrier to form a bond, but no diffusion to any extent away from the area of the bond. The barrier zirconium is of the order preferably about 1 to 30 percent, and more preferably 5 to 15 percent of the total thickness of the tubing. In a particularly preferred embodiment, a thickness of 10 percent of the tubing bonded to the alloy tube of a zirconium alloy provides good stress reduction and a barrier effect sufficient to prevent failures in the tubing. III. MANUFACTURE OF THE TUBING The size of the precipitate can be controlled by various manufacturing processes. Initially, the precipitate size is governed essentially by the cooling or quenching rate from the beta phase. The beta phase refers to the body-centered cubic crystal lattice structure of crystalline zirconium and Zircaloy that stable at higher temperatures (i.e. forms above about 960° C for Zircaloy-2). A different phase, the alpha phase, is a close-packed hexagonal crystal lattice structure of zirconium and Zircaloy that is stable at lower temperatures. Between about 825° C and 960° C, the alpha and beta phases coexist in Zircaloys. Rapid quenching rates from the beta phase (e.g. faster than about 50° C per second) give smaller precipitates, while slower cooling rates give larger precipitates. The initial precipitate sizes (obtained by quenching from the beta phase) can be altered somewhat by later heat treatments such as annealing at a high temperature (e.g. greater than about 621° C). This allows some of the nickel, iron and chrome components of the Zircaloy matrix phase and the smaller precipitates to diffuse to larger precipitates, causing the precipitates to coarsen. A guideline widely applicable to various processes is provided by the "accumulated normalized annealing time" defined in the article of F. Garzarolli, et al ., "Progress in the Knowledge of Nodular Corrosion", Zirconium in the Nuclear Industry , ASTM STP939, pp. 417-430 (1987), which is incorporated herein by reference. Preferably, to ensure sufficiently coarse precipitates, the accumulated normalized annealing time should be greater than about 10 -17 hours. To obtain the nonuniform precipitate distribution of the present invention in which fine precipitates are confined to the outer regions of the tubing, the outer and inner regions of the tubing have to be maintained at different temperatures during at least one process step. This is conveniently accomplished by known induction heat treatment procedures such as those described in U.S. Patent No. 4,576,654. Generally, the tubing is heated in an induction coil while flowing cooling water through the center of the tube. This raises the temperature of the outer region sufficiently to transform it to the beta phase, while holding the temperature of the inner region to a lower level, thus preserving the coarse precipitate structure. The tube is then rapidly quenched to produce small precipitates in the outer region only. Regarding the cooling of the tube, any fluid which is generally intert with respect to the zirconium alloy or barrier material can be used. For example, a gas coolant, water, or even steam can be used in such a process. Of course, by using such techniques a metallurgical gradient or transition region will be present between the outer fine precipitate region of the tubing and the bulk coarse precipitate region. In this transition region, the precipitate size will vary between the desired fine and coarse sizes. In general, it will be a relatively sharp boundary. As used herein, the transition region is considered to be part of the outer region. Typically, it will occupy no more than about 25% of the outer region, but this is by no means a required value. In some cases, the transition region will occupy a much as 5 to 10% of the entire tube wall. To obtain the final tubing of the necessary dimensions, various other manufacturing steps such as cold-working, extruding, heat treating, and annealing may be employed. The equipment and operating conditions necessary to carry out these various steps will be readily apparent to those of skill in the art, and are described in U.S. Patent 5,437,747. European patent application EP-A-0 622 470 entitled METHOD OF FABRICATING ZIRCALOY TUBING HAVING HIGH RESISTANCE TO CRACK PROPAGATION. In a preferred embodiment, a billet of zirconium alloy is beta quenched from 1000° C to about 700° C by immersion in a tank of water. Next, the tube is extruded with the tube temperature being at about 570° C by putting the tube through a set of tapered dies under high pressure. The extruded product is referred to as a "tubeshell" which is available in specified dimensions from various vendors such as Teledyne Wahchang (Albany, Oregon USA), Western Zirconium (A Westinghouse company of Ogden, Utah), and Cezus (France). Next, a first pass cold work to 70% is performed as in the known processes followed by an anneal at a relatively high temperature (e.g. 650° C for four hours). At this point, a heat treatment is performed to regain the smaller precipitates on the outside of the tube. This treatment is performed at 1045° C (in the pure beta phase). An induction coil rapidly heats the outer 15% of the tube to the desired temperature and then shuts off while water is flowing through the tube interior. This allows the tube to rapidly cool (sometimes within a matter of 2 seconds). The penetration of the induction coil energy can be tuned by adjusting the induction coil frequency, the induction coil energy, the speed at which the tube moves through the induction coil, and the water temperature (flow rate). One of skill in the art will appreciate how to adjust these conditions to achieve the type of heat treatment that will form small precipitates at the outer 15% of the tube. Further details can be found in U.S. patent No. 4,576,654 to Eddens . The resulting tube will have good nodular corrosion resistance while retaining coarse precipitates in the inner regions. Next, a second pass cold work to 70% is performed followed by annealing at 650° C for 2 hours. A third pass cold work and a recrystallization or stress release anneal are performed under the same conditions as the known processes. At this point the tubing is suitable (with only minor modifications and testing) for use in a fuel rod. If the tubing of this invention is to contain a barrier layer, the fabrication process will have at least one additional step. Details of such a step are known in the art and are provided in, for example, U.S. patent No. 4,894,203. Usually, the barrier layer is bonded to the tubing as liner during an extrusion step. Other steps in the process are performed as described above. IV. CONCLUSION Although the foregoing invention has been described in some detail for purposes of clarity of understanding, it will be apparent that certain changes and modifications may be practiced within the scope of the appended claims. For instance, although the specification has described preferred zirconium alloy tubes, other shapes may be used as well. For example, plates and metal sections of other shapes may also be used. The Zircaloys described above are examples of alloys that can advantageously be used as tubing in the present invention. Some other zirconium-based alloys as well as certain titanium-based alloys and other metal alloys having similar structures can in many instances also employ the precipitate microstructure of this invention to resist corrosion and crack propagation. I prefer the use of this invention with cladding in the reactor. The reader will understand that the metallurgy herein can be used in other reactor parts. For example, the Zirconium alloy composition here taught may be used with water rods, spacers, channels and other Zirconium alloy structures and their equivalent within the reactor.

Claims: A zirconium alloy cladding tube (17) with a metallurgical gradient, said cladding tube comprising a corrosion resistant outer circumferential tube region (33) comprising relatively fine and more densely distributed precipitates (108) and a crack resistant interior circumferential tube region (30) CHARACTERIZED BY said interior circumferential tube region (30) comprising coarser and less densely distributed precipitates (102) having an average diameter of at least 0.15 µm (microns) produced by coarse anneal at temperature of greater than 621°C. The zirconium alloy cladding tube of claim 1 wherein the interior circumferential tube region occupies between about 80 and 95% of the tube cross-section. The zirconium alloy cladding tube of claim 1 further comprising a barrier layer (31). The zirconium alloy cladding tube of claim 1 wherein the coarser precipitates (102) have an average diameter of between about 0.15 and 2 µm (microns). The zirconium alloy cladding tube of claim 1 wherein the zirconium alloy comprises at least about 98% zirconium by weight, between 0.06 and 0.25% iron by weight, between 0.03 and 0.1% nickel by weight, and between about 0.8 and 1.7% tin by weight. The zirconium alloy cladding tube of claim 1 wherein the outer circumferential tube region (33) comprising relatively fine and more densely distributed precipitates and the interior circumferential tube region (30) comprising relatively coarser and less densely distributed precipitates are produced by one or more anneals that together have an accumulated normalized annealing time of at least 10 -17 hours.

Inventors: ADAMSON RONALD BERT, POTTS GERALD ALLEN, ADAMSON, RONALD BERT, POTTS, GERALD ALLEN
Applicants: GEN ELECTRIC, GENERAL ELECTRIC COMPANY
Application year: 1994
Main classification: G21C3
IPC codes: C22C16, C22F1, G21C3